Long-Term Radioactivity Release from Solidified High-Level Waste - Part II Parametric Study of Waste form Properties, Temperature and Time

1982 ◽  
Vol 15 ◽  
Author(s):  
Friedrich K. Altenhein ◽  
Werner Lutze ◽  
Rodney C. Ewing
Keyword(s):  
Parametric Study ◽  
Computer Code ◽  
Salt Dome ◽  
Waste Form ◽  
Glass Block ◽  
High Level Waste ◽  
Surface Layers ◽  
History Of ◽  
High Level

The computer code QTERM has been used to calculate the total released activity from a single glass block when in contact with brine in a salt dome repository as a function of: (1) waste form properties, (2) leaching mechanisms, (3) retention or precipitation of specific radionuclides in surface layers, (4) thermal history of the repository and (5) decreasing activity as a function of time.

Assessing the durability of nuclear glass with respect to silica controlling processes in a clayey underground disposal

2006 ◽  
Vol 932 ◽  
Author(s):  
Laurent De Windt ◽  
Stéphanie Leclercq ◽  
Jan van der Lee
Keyword(s):  
Host Rock ◽  
Reactive Transport ◽  
Transport Model ◽  
Glass Block ◽  
High Level Waste ◽  
Residual Rate ◽  
Glass Dissolution ◽  
Underground Disposal ◽  
High Level

ABSTRACTThe long-term behaviour of vitrified high-level waste in an underground clay repository was assessed by using the reactive transport model HYTEC with respect to silica diffusion, sorption and precipitation processes. Special attention was given to the chemical interactions between glass, corroded steel and the host-rock considering realistic time scale and repository design. A kinetic and congruent dissolution law of R7T7 nuclear glass was used assuming a first-order dissolution rate, which is chemistry dependent, as well as a long-term residual rate. Without silica sorption and precipitation, glass dissolution is diffusion-driven and the fraction of altered glass after 100,000 years ranges from 5% to 50% depending on the fracturation degree of the glass block. Corrosion products may limit glass dissolution by controlling silica diffusion, whereas silica sorption on such products has almost no effect on glass durability. Within the clayey host-rock, precipitation of silicate minerals such as chalcedony may affect glass durability much more significantly than sorption. In that case, however, a concomitant porosity drop is predicted that could progressively reduce silica diffusion and subsequent glass alteration.

Long-Term Radioactivity Release From Solidified High-Level Waste -Part I: An Approach To Evaluating Experimental Data

1981 ◽  
Vol 11 ◽  
Author(s):  
Friedrich K. Altenhein ◽  
Werner Lutze ◽  
Rodney C. Ewing
Keyword(s):  
Source Term ◽  
Elevated Temperatures ◽  
Relative Sensitivity ◽  
Reaction Rates ◽  
Waste Form ◽  
High Level Waste ◽  
Leach Rate ◽  
Leaching Experiments ◽  
High Level

Safety and risk analyses for the isolation of radioactive waste in a repository must begin with a source term to quantify the amount of radioactivity released from the waste form under a specific set of conditions. The interaction of the waste form with aqueous solutions is the most important mechanism to consider, as any material released may be dissolved and reach the biosphere. In this regard the behaviour of heat generating high-level waste is of particular importance, because reaction rates are higher at elevated temperatures. A long-term leach rate was derived from previous and continuing experimental work. The purpose of this paper is not to describe the “real case” release but rather to provide guidelines for the design of leaching experiments and determine the required precision for the data. This can be derived from the relative sensitivity of extrapolated leach rates for various parameters measured in laboratory experiments.

New results of Boda Claystone research: Genesis, mineralogy, geochemistry, petrophysics

2018 ◽  
Vol 482 (1) ◽  
pp. 75-92 ◽  
Author(s):  
Ferenc Fedor ◽  
Zoltán Máthé ◽  
Péter Ács ◽  
Péter Koroncz
Keyword(s):  
High Level Waste ◽  
Swelling Clays ◽  
Theoretical Approaches ◽  
Hydrodynamic Behaviour ◽  
Geological Formation ◽  
Porosity And Permeability ◽  
History Of ◽  
Surrounding Areas ◽  
High Level ◽  
Low Porosity

AbstractBoda Claystone is a very tight clayey rock with extreme low porosity and permeability, nano-size pores and small amounts of swelling clays. Due to this character it is ideal as a potential host rock for research into the possibilities of high-level waste deposition in geological formation. Though the research started more than 30 years ago, the genesis, the geotectonic history of the Boda Claystone Formation (BCF) and the geology of surrounding areas has only been sketched out recently. On the basis of research of the past few years the process of sedimentation of different blocks was able to be reconstructed. Equipment and methodological developments were needed for the investigation of reservoir geological and hydrodynamic behaviour of this rock, which began in the early 2000s. Based on them the pore structure and reservoir could be characterized in detail. Only theoretical approaches were available for the chemical composition of free porewater. Traditional water-extracting methods were not adaptable because of excessively low porosity and nano-scale pore size distribution. Hence, new ways have to be found for getting enough water for analysis. These new results of BCF research help to prepare more sophisticated and directed experiments, in which there is a great interest internationally.

scholarly journals Spent Fuel as a Waste Form - Data Needs to Allow Long Term Performance Assessment Under Repository Disposal Conditions.

1986 ◽  
Vol 84 ◽  
Author(s):  
V. M. Oversby
Keyword(s):  
Performance Assessment ◽  
Waste Form ◽  
High Level Waste ◽  
Detailed Knowledge ◽  
Spent Fuel ◽  
Expected Performance ◽  
High Level ◽  
Term Performance ◽  
Physical And Chemical ◽  
Waste Repositories

AbstractPerformance assessment calculations are required for high level waste repositories for a period of 10,000 years under NRC and EPA regulations. In addition, the Siting Guidelines (IOCFR960) require a comparison of sites following site characterization and prior to final site selection to be made over a 100,000 year period. In order to perform the required calculations, a detailed knowledge of the physical and chemical processes that affect waste form performance will be needed for each site. While bounding calculations might be sufficient to show compliance with the requirements of IOCFR60 and 40CFRI91, the site comparison for 100,000 years will need to be based on expected performance under site specific conditions. The only case where detailed knowledge of waste form characteristics in the repository would not be needed would be where radionuclide travel times to the accessible environment can be shown to exceed 100,000 years. This paper will review the factors that affect the release of radionuclides from spemt fuel under repository conditions, summarize our present state of knowledge, and suggest areas where more work is needed in order to support the performance assessment calculations.

Lessons Learned Siting and Successfully Processing U.S. DOE Radioactive Wastes Using a High Throughput Vitrification Process

2003 ◽  
Author(s):  
Robert E. Prince ◽  
Bradley W. Bowan
Keyword(s):  
The United States ◽  
Lessons Learned ◽  
Volume Reduction ◽  
Waste Form ◽  
High Level Waste ◽  
Hanford Site ◽  
High Level ◽  
Low Activity ◽  
The U.S ◽  
Ceramic Melter

This paper describes actual experience applying a technology to achieve volume reduction while producing a stable waste form for low and intermediate level liquid (L/ILW) wastes, and the L/ILW fraction produced from pre-processing of high level wastes. The chief process addressed will be vitrification. The joule-heated ceramic melter vitrification process has been used successfully on a number of waste streams produced by the U.S. Department of Energy (DOE). This paper will address lessons learned in achieving dramatic improvements in process throughput, based on actual pilot and full-scale waste processing experience. Since 1991, Duratek, Inc., and its long-term research partner, the Vitreous State Laboratory of The Catholic University of America, have worked to continuously improve joule heated ceramic melter vitrification technology in support of waste stabilization and disposition in the United States. From 1993 to 1998, under contact to the DOE, the team designed, built, and operated a joule-heated melter (the DuraMelterTM) to process liquid mixed (hazardous/low activity) waste material at the Savannah River Site (SRS) in South Carolina. This melter produced 1,000,000 kilograms of vitrified waste, achieving a volume reduction of approximately 70 percent and ultimately producing a waste form that the U.S. Environmental Protection Agency (EPA) delisted for its hazardous classification. The team built upon its SRS M Area experience to produce state-of-the-art melter technology that will be used at the DOE’s Hanford site in Richland, Washington. Since 1998, the DuraMelterTM has been the reference vitrification technology for processing both the high level waste (HLW) and low activity waste (LAW) fractions of liquid HLW waste from the U.S. DOE’s Hanford site. Process innovations have doubled the throughput and enhanced the ability to handle problem constituents in LAW. This paper provides lessons learned from the operation and testing of two facilities that provide the technology for a vitrification system that will be used in the stabilization of the low level fraction of Hanford’s high level tank wastes.

scholarly journals Effect of Fe2O3 on the Immobilization of High-Level Waste with Magnesium Potassium Phosphate Ceramic

2019 ◽  
Vol 2019 ◽  
pp. 1-10
Author(s):  
Hailin Yang ◽  
Mingjiao Fu ◽  
Bobo Wu ◽  
Ying Zhang ◽  
Ruhua Ma ◽  
...  
Keyword(s):  
Sintering Temperature ◽  
Potassium Phosphate ◽  
High Level Waste ◽  
Radioactive Nuclide ◽  
Hydration Reaction ◽  
Sintering Process ◽  
Positive Role ◽  
Ceramic Structure ◽  
High Level

For the proposed novel procedure of immobilizing HLW with magnesium potassium phosphate cement (MKPC), Fe2O3 was added as a modifying agent to verify its effect on the solidification form and the immobilization of the radioactive nuclide. The results show that Fe2O3 is inert during the hydration reaction. It slows down the hydration reaction and lowers the heat release rate of the MKPC system, leading to a 3°C-5°C drop in the mixture temperature during hydration. Early comprehensive strength of Fe2O3 containing samples decreased slightly while the long-term strength remained unchanged. For the sintering process, Fe2O3 played a positive role, lowering the melting point and aiding the formation of ceramic structure. CsFe(PO4)2, or CsFePO4, was generated by sintering at 900°C. These products together with the ceramic structure and absorption benefit the immobilization of Cs+. The optimal sintering temperature for heat treatment is 900°C; it makes the solidification form a fired ceramic-like structure.

Flow Barrier System for Long-Term High-Level-Waste Isolation: Experimental Results

1998 ◽  
Vol 124 (1) ◽  
pp. 88-100 ◽  
Author(s):  
James L. Conca ◽  
Michael J. Apted ◽  
Wei Zhou ◽  
Randolph C. Arthur ◽  
John H. Kessler
Keyword(s):  
Experimental Results ◽  
High Level Waste ◽  
High Level
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